Accident management following loss of residual heat removal during mid-loop operation in a Westinghouse two-loop PWR

Birchley, J.; Haste, T.; Richner, M.

doi:10.1016/j.nucengdes.2008.02.009

Cited by 14 publications

(4 citation statements)

References 5 publications

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“…High pressures of 4 to 10 MPa can cause reflux condensation. This can happen because of a LOCA during normal plant operation [133] or at containment pressures resulting from the loss of the RHRS during midloop inventory operations [134].…”

Section: Separate Effect Experimentsmentioning

confidence: 99%

Steam Condensation Heat Transfer in the Presence of Noncondensable Gases (NCGs) in Nuclear Power Plants (NPPs): A Comprehensive Review of Fundamentals, Current Status, and Prospects for Future Research

Albdour,

Addad,

Alyammahi

et al. 2024

International Journal of Energy Research

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Efficient steam condensation is crucial for safe nuclear power plant (NPP) operations, preventing pressure buildup, overheating, and the release of radioactive materials. However, the presence of noncondensable gases (NCGs), such as air, nitrogen, hydrogen, or helium, can hinder the condensation process by creating a thermal resistance layer that impedes steam diffusion and condensation on the system’s surface. Maximizing the efficiency of steam condensation requires a thorough grasp of the fundamental processes, theories, advancements, and technical hurdles. Therefore, this work thoroughly addresses these needs, with a particular emphasis on addressing the challenges posed by NCGs by dividing the work into four thematic areas. The first theme relates to a comprehensive examination of pure steam condensation phenomena, which includes an exploration of familiar condensation scenarios and various film condensation types. The second theme examines condensation in the presence of NCGs, their mixture properties, and related theories and modelling of heat and mass transfer. The third theme investigates condensation in NPP by exploring passive cooling systems and condensation phenomena under both natural and forced convection conditions during nuclear accidents, the origin of NCGs in NPP and their transportation aspects. This is followed by experimental work related to condensation scenarios and scale. Finally, the last theme looks upon the recent advancements in computational fluid dynamics (CFD) modelling of wall condensation, system analysis codes coupling with CFD, and the implementation of machine learning (ML) for predicting the condensation HTC. By bridging the gap between fundamental knowledge and practical applications, the four thematic areas presented in this work are aimed at providing a comprehensive foundation for researchers and experts in the field of steam condensation when NCGs are involved. The ultimate objective is to bolster the safety and efficacy of NPP operations by understanding the heat and mass transfer mechanisms while mitigating the risk of catastrophic events.

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Section: Separate Effect Experimentsmentioning

confidence: 99%

Steam Condensation Heat Transfer in the Presence of Noncondensable Gases (NCGs) in Nuclear Power Plants (NPPs): A Comprehensive Review of Fundamentals, Current Status, and Prospects for Future Research

Albdour,

Addad,

Alyammahi

et al. 2024

International Journal of Energy Research

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“…This code is to treat the entire spectrum of severe accident phenomena, including reactor coolant system and containment thermal-hydraulic response, core heat-up, degradation and relocation, and fission product release and transport, in a unified framework for both PWR and BWR [13,14]. MELCOR code is mostly used by regulatory body and academic researchers to evaluate strategies for severe accident mitigation and to simulate detailed features of severe accident sequences [15][16][17][18][19][20]. In Korea, for a successful severe accident management, an AtomCARE system has been developed for collection of SAMG variables connecting plant monitoring program with MELCOR [21].…”

Section: Melcor Descriptionmentioning

confidence: 99%

Validation of RCS depressurization strategy and core coolability map for independent scenarios of SBLOCA, SBO, and TLOFW

Lee

Kim

et al. 2013

Journal of Nuclear Science and Technology

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Severe accident analysis for Small Break (SB), Middle Break (MB), and Large Break (LB) Loss-Of-Coolant Accident (LOCA), Station Black Out (SBO), Total Loss-Of-Feed-Water (TLOFW) was performed and effectiveness of Reactor Coolant System (RCS) depressurization strategies of OPR1000 was analyzed using MELCOR 1.8.6 code. Required injection flow rate has been derived using Core Exit Temperature (CET) information obtained from MELCOR calculation and a simple model and corresponding coolability map have been suggested to assist effective operator action. The depressurization strategies using secondary Atmospheric Dump Valve (ADV) for SBLOCA, pressurizer Safety Depressurization System (SDS) for SBO and TLOFW were introduced in 5 min since the initiation of Severe Accident Management Guidance (SAMG). Respective mitigation strategy employed leads to significant delay of the reactor pressure vessel failure and RCS pressure at Reactor Pressure Vessel (RPV) failure was lower than the SAMG target pressure of 2.86 MPa. Thus, possibility of High Pressure Melt Ejection (HPME) and impair of containment building is expected to avoid effectively. Using CET information obtained from MELCOR calculation, a simple model and a coolability map for the required injection flow rate were developed for recovery of core coolability. It is suggested that the coolability map based on MELCOR calculation results may provide decisive and intuitive information to operators for more effective safety management.

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“…As a result, a possibly significant passive residual decay heat removal from the reactor's primary coolant system is established [1][2][3] . Reflux condensation can occur at high pressures 4-10 MPa, due to a loss of coolant accident (LOCA) during normal plant operation [4][5][6][7][8] or at containment pressures due to loss of residual heat removal (RHR) system during a mid-loop inventory operation [9][10][11] .…”

Section: Introductionmentioning

confidence: 99%

The use of a novel gradient heat flux sensor for characterization of reflux condensation

Janasz

Prasser

Suckow

et al. 2022

Nuclear Engineering and Design

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Accident management following loss of residual heat removal during mid-loop operation in a Westinghouse two-loop PWR

Cited by 14 publications

References 5 publications

Steam Condensation Heat Transfer in the Presence of Noncondensable Gases (NCGs) in Nuclear Power Plants (NPPs): A Comprehensive Review of Fundamentals, Current Status, and Prospects for Future Research

Steam Condensation Heat Transfer in the Presence of Noncondensable Gases (NCGs) in Nuclear Power Plants (NPPs): A Comprehensive Review of Fundamentals, Current Status, and Prospects for Future Research

Validation of RCS depressurization strategy and core coolability map for independent scenarios of SBLOCA, SBO, and TLOFW

The use of a novel gradient heat flux sensor for characterization of reflux condensation

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